Plasma Vessel Shield Layout for Tritium Breeding and Neutron Protection
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Solution Overview
Problem
Existing neutron shields in plasma vessels for nuclear fusion reactors are inefficient in areas where the space between plasma confinement coils is limited, leading to unnecessary material usage and reduced efficiency, as they do not adequately address neutron penetration and fuel breeding needs.
Innovation Solution
A plasma vessel design with a varying ratio of fusion fuel breeding blanket to neutron shield thickness as a function of poloidal and toroidal angles, optimizing shielding and breeding based on available space and neutron flux, using materials like tungsten for enhanced neutron attenuation.
Engineering Contradictions & Design Principles
Engineering Contradiction Analysis
1Reliability
If the thickness of the neutron shield is uniformly increased throughout the plasma vessel, then the shielding effectiveness against high-energy neutrons is improved, but the material usage increases and the space for fusion fuel breeding is reduced
Solution Approach 1:
The patent applies local quality by varying the neutron shield thickness according to the local neutron flux distribution. The shield thickness is increased in regions with high neutron flux (such as near the plasma core) and reduced in regions with lower neutron flux (such as near the plasma confinement coils), thereby optimizing the balance between shielding effectiveness and material usage.
Solution Approach 2:
The patent segments the neutron shield into multiple regions with different thicknesses based on the spatial distribution of neutron flux. This segmentation allows each region to have the appropriate shield thickness for its specific neutron exposure level, avoiding uniform over-shielding throughout the entire vessel.
2Productivity
If the thickness of the fusion fuel breeding blanket is increased, then the tritium breeding capability is improved, but the shielding effectiveness against high-energy neutrons is reduced
Solution Approach 1:
The patent applies local quality by varying the breeding blanket thickness in different regions of the plasma vessel. The breeding blanket is made thicker in regions where neutron flux is lower (to enhance tritium breeding) and thinner in regions where neutron flux is high (to maintain adequate shielding), thereby optimizing both tritium breeding and neutron protection simultaneously.
3Volume of moving object
If the plasma vessel thickness is adapted to available space, then the space utilization is improved, but the shielding effectiveness in regions with limited space is reduced
Solution Approach 1:
The patent applies local quality by adapting the shield and breeding blanket thickness to the locally available space. In regions with limited space between the plasma confinement and coils, the design optimizes the distribution of shield and breeding blanket thicknesses to provide adequate neutron protection while maximizing the use of available volume for fusion fuel breeding.
Applied Scientific Principles
This section explains which scientific principles are used to turn an abstract innovation direction into a practical engineering solution.
Function Achieved in This Case
Maximizes tritium breeding and thermal energy extraction while protecting plasma confinement coils from high-energy neutrons, enhancing the safety and efficiency of nuclear fusion reactors.
Implementation Method 1
The primary function of a neutron shield is to absorb and attenuate high-energy neutrons produced during the fusion process
Implementation Method 2
In nuclear fusion, isotopes like deuterium and tritium combine to form helium, releasing a significant number of high-energy neutrons
Implementation Method 3
By absorbing these neutrons, the shield contributes to the moderation and thermalization of neutrons, allowing for the recovery of their kinetic energy
Implementation Method 4
Stellarators are developed for magnetic confinement of plasmas to provide energy based on fusion reactions
Implementation Method 5
The plasma is confined in a plasma vessel having a vacuum chamber, comprising a breeding blanket for providing the necessary fuel for the fusion reaction
Implementation Method 6
The superconductor material allows a positive energy balance of the stellarator as the current in the cable has no electrical resistance, thereby avoiding ohmic losses of the cable during operation
Data Source
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AI summary
The invention relates to a plasma vessel for a nuclear fusion reactor, comprising a neutron shield and a fusion fuel breeding blanket, the plasma vessel enclosing a plasma confinement volume. A ratio of a thickness of the fusion fuel breeding blanket to a thickness of the neutron shield varies as a function of a poloidal angle and / or of a toroidal angle from o° to 360° of the plasma vessel by at least 5 %. The invention also relates to a stellarator comprising the plasma vessel, and a computer implemented design method for optimizing a plasma vessel for a nuclear fusion reactor as well as a corresponding computer program and computer device.